ML19035A760

Attachment for Amendment 153 - LAR-18-024

ML19035A760
Person / Time
Site:	Vogtle
Issue date:	02/22/2019
From:	NRC/NRO/DLSE/LB2
To:	Southern Nuclear Operating Co
GLEAVES B/415-5848
Shared Package
ML19035A755	List: ML19035A756 ML19035A757 ML19035A758 ML19035A759 ML19035A760 ML19035A761 ML19035A762 ML19035A763
References
EPID L-2018-LLA-0237
Download: ML19035A760 (4)

Text

ATTACHMENT TO LICENSE AMENDMENT NO. 153 TO FACILITY COMBINED LICENSE NO. NPF-92 DOCKET NO.52-026 Replace the following pages of the Facility Combined License No. NPF-92 with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Facility Combined License No. NPF-92 REMOVE INSERT 7 7 9 9 11 11

(7) Reporting Requirements (a) Within 30 days of a change to the initial test program described in UFSAR Section 14, Initial Test Program, made in accordance with 10 CFR 50.59 or in accordance with 10 CFR Part 52, Appendix D, Section VIII, Processes for Changes and Departures, SNC shall report the change to the Director of NRO, or the Directors designee, in accordance with 10 CFR 50.59(d).

(b) SNC shall report any violation of a requirement in Section 2.D.(3),

Section 2.D.(4), Section 2.D.(5), and Section 2.D.(6) of this license within 24 hours1 days <br />0.143 weeks <br />0.0329 months <br />. Initial notification shall be made to the NRC Operations Center in accordance with 10 CFR 50.72, with written follow up in accordance with 10 CFR 50.73.

(8) Incorporation The Technical Specifications, Environmental Protection Plan, and ITAAC in Appendices A, B, and C, respectively of this license, as revised through Amendment No. 153, are hereby incorporated into this license.

(9) Technical Specifications The technical specifications in Appendix A to this license become effective upon a Commission finding that the acceptance criteria in this license (ITAAC) are met in accordance with 10 CFR 52.103(g).

(10) Operational Program Implementation SNC shall implement the programs or portions of programs identified below, on or before the date SNC achieves the following milestones:

(a) Environmental Qualification Program implemented before initial fuel load; (b) Reactor Vessel Material Surveillance Program implemented before initial criticality; (c) Preservice Testing Program implemented before initial fuel load; (d) Containment Leakage Rate Testing Program implemented before initial fuel load; (e) Fire Protection Program

1. The fire protection measures in accordance with Regulatory Guide (RG) 1.189 for designated storage building areas (including adjacent fire areas that could affect the storage area) implemented before initial receipt 7 Amendment No. 153

(m) Initial Test Program (ITP)

1. Preoperational Test Program implemented before the first preoperational test; and

2. Startup Test Program implemented before initial fuel load; (n) Special Nuclear Material Control and Accounting Program implemented before initial receipt of special nuclear material; and (o) Special Nuclear Material Physical Protection Program implemented before initial receipt of special nuclear material on site.

(11) Operational Program Implementation Schedule No later than 12 months after issuance of the COL, SNC shall submit to the Director of NRO, or the Directors designee, a schedule for implementation of the operational programs listed in UFSAR Table 13.4-201, including the associated estimated date for initial loading of fuel. The schedule shall be updated every 6 months until 12 months before scheduled fuel loading, and every month thereafter until all the operational programs listed in UFSAR Table 13.4-201 have been fully implemented.

(12) Site- and Unit-specific Conditions (a) SNC shall either remove and replace, or shall improve, the soils directly above the blue bluff marl for soils under or adjacent to Seismic Category I structures, to eliminate any liquefaction potential.

(b) Before commencing installation of individual piping segments and connected components in their final locations, SNC shall complete the as-designed pipe rupture hazards analysis for compartments (rooms) containing those segments in accordance with the criteria outlined in the UFSAR Sections 3.6.1.3.2 and 3.6.2.5, and shall inform the Director of NRO, or the Directors designee, in writing, upon the completion of this analysis and the availability of the as-designed pipe rupture hazards analysis reports.

(c) Before commencing installation of individual piping segments, identified in UFSAR Section 3.9.8.7, and connected components in their final locations in the facility, SNC shall complete the analysis of the as-designed individual piping segments and shall inform the Director of NRO, or the Directors 9 Amendment No. 153

power calorimetric uncertainty instrumentation (before initial fuel load);

7. The site-specific severe accident management guidelines (before startup testing);

8. The operational and programmatic elements of the mitigative strategies for responding to circumstances associated with loss of large areas of the plant due to explosions or fire developed in accordance with 10 CFR 50.54(hh)(2) (before initial fuel load); and

9. The ITP procedures identified in U FSAR Section 14.2.3:

a. administrative manual (before the first preoperational test)

b. preoperational testing (before scheduled performance)

c. startup testing (before initial fuel load)

(g) Before initial fuel load, SNC shall:

1. Update the seismic interaction analysis in UFSAR Section 3.7.5.3 to reflect as-built information, which must be based on as-procured data, as well as the as-constructed condition;

2. Reconcile the seismic analyses described in Section 3.7.2 of the UFSAR, to account for detailed design changes, including, but not limited to, those due to as-procured or as-built changes in component mass, center of gravity, and support configuration based on as-procured equipment information;

3. Calculate the instrumentation uncertainties of the actual plant operating instrumentation to confirm that either the design limit departure from nucleate boiling ratio (DNBR) values remain valid or that the safety analysis minimum DNBR bounds the new design limit DNBR values plus DNBR penalties;

4. Update the pressure temperature (P-T) limits using the pressure temperature limits report (PTLR) methodologies approved in the UFSAR, using the plant-specific material properties or confirm that the reactor vessel material properties meet the specifications of and use the Westinghouse generic PTLR curves;

5. Verify that plant-specific belt line material properties are consistent with the properties given in UFSAR Section 5.3.3.1 and Tables 5.3-1 and 5.3-3. The verification must include a pressurized thermal shock (PTS) evaluation based on as-procured reactor vessel material data and the projected neutron fluence for the 11 Amendment No. 153